Generic Aging Lessons Learned for Subsequent …NUREG–2191, Vol. 1 Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report Final Report Office of Nuclear Reactor

NUREG–2191, Vol. 1

Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report

Final Report

Office of Nuclear Reactor Regulation

NUREG–2191, Vol. 1

Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report

Final Report

Office of Nuclear Reactor Regulation

Manuscript Completed: May 2017 Date Published: July 2017

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ABSTRACT

This document provides guidance on the content of applications for renewal of the initial renewed operating license. The initial renewed operating license is the first renewed license issued under Title 10 of the Code of Federal Regulations (10 CFR) Part 54, “Requirements for Renewal of Operating Licenses for Nuclear Power Plants,” after either supersession or the expiration of the original operating license issued under either 10 CFR Part 50 or Part 52 following the completion of construction under a construction permit issued under Part 50, or a combined license issued under Part 52. In this guidance document, the renewal of the initial renewed operating license is referred to as “subsequent license renewal” (SLR). NUREG–2191, “Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report,” provides guidance for SLR applicants. The GALL-SLR Report contains the U.S. Nuclear Regulatory Commission (NRC) staff’s generic evaluation of plant aging management programs (AMPs) and establishes the technical basis for their adequacy. The GALL-SLR Report contains recommendations on specific areas for which existing AMPs should be augmented for SLR. An applicant may reference this report in an SLR application to demonstrate that the AMPs at the applicant’s facility correspond to those described in the GALL-SLR Report. If an applicant credits an AMP in the GALL-SLR Report, it is incumbent on the applicant to ensure that the conditions and operating experience (OE) at the plant are bounded by the conditions and OE for which the GALL-SLR Report program was evaluated. If these bounding conditions are not met, it is incumbent on the applicant to address any additional aging effects and augment the AMPs for SLR. For AMPs that are based on the GALL-SLR Report, the NRC staff will review and verify whether the applicant’s AMPs are consistent with those described in the GALL-SLR Report, including applicable plant conditions and OE. The focus of the NRC staff’s review of an SLR application is on those AMPs that an applicant has enhanced to be consistent with the GALL-SLR Report, those AMPs for which the applicant has taken an exception to the program described in the GALL-SLR Report, and plant-specific AMPs not described in the GALL-SLR Report.

This document is a companion document to NUREG–2192, “Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants,” (SRP-SLR) that provides guidance to NRC staff on the review of SLR applications. The guidance in this document is for the use of future applicants for SLR. The NRC does not intend to impose the guidance in this document on current holders of an initial operating license. However, this document encompasses all of the guidance applicable to initial license renewal. Accordingly, both current holders of initial operating licenses as well as future applicants for initial license renewal may voluntarily choose to reference an AMP in the GALL-SLR Report in their applications. However, such applicants should inform the NRC that they plan to demonstrate consistency with the GALL-SLR Report.

Both the GALL-SLR Report and the SRP-SLR were published for public comment in December 2015, with the comment period ending February 29, 2016. The staff received over 300 pages of comments from interested stakeholders. These comments were reviewed and dispositioned by the staff. The disposition of these comments and the technical bases for the staffs’ agreement or disagreement with these comments will be published shortly in a NUREG. The staff will also publish a second NUREG that will document all the technical changes made to the license renewal guidance documents for first license renewal (i.e., for operation from 40 years to 60 years), along with the technical bases for these changes.

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TABLE OF CONTENTS

Section Page

ABSTRACT ........................................................................................................................................ iii LIST OF TABLES .............................................................................................................................. xi LIST OF CONTRIBUTORS ..............................................................................................................xv ABBREVIATIONS ........................................................................................................................... xix INTRODUCTION ............................................................................................................................. xxv BACKGROUND ............................................................................................................................ xxvii OVERVIEW OF THE GENERIC AGING LESSONS LEARNED FOR SUBSEQUENT LICENSE RENEWAL (GALL-SLR) REPORT EVALUATION PROCESS .................................. xxxi EXPLANATION OF THE USE OF MULTIPLE AGING MANAGEMENT PROGRAMS IN AGING MANAGEMENT REVIEW ITEMS ............................................................................... xxxv REFERENCES ............................................................................................................................ xxxvii GUIDANCE ON USE OF LATER EDITIONS/REVISIONS OF VARIOUS INDUSTRY DOCUMENTS ......................................................................................... xxxix APPLICATION OF THE GENERIC AGING LESSONS LEARNED FOR SUBSEQUENT LICENSE RENEWAL (GALL-SLR) REPORT ................................................................................. xli

I APPLICATION OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILER AND PRESSURE VESSEL CODE ...................................................................... l-1

II CONTAINMENT STRUCTURES ....................................................................................... II-1 II PRESSURIZED WATER REACTOR CONTAINMENTS ................................... II-3 A1 CONCRETE CONTAINMENTS (REINFORCED AND PRESTRESSED) ....................................................... II A1-1 A2 STEEL CONTAINMENTS ............................................................................. II A2-1 A3 COMMON COMPONENTS ........................................................................... II A3-1 II BOILING WATER REACTOR CONTAINMENTS ........................................... II B-1 B1 MARK I CONTAINMENTS ............................................................................ II B1-1 B2 MARK II CONTAINMENTS ........................................................................... II B2-1 II MARK III CONTAINMENTS .............................................................................. II B3 B3 MARK III CONTAINMENTS .......................................................................... II B3-1 B4 COMMON COMPONENTS ........................................................................... II B4-1

III STRUCTURES AND COMPONENT SUPPORTS ........................................................ III-1 III SAFETY-RELATED AND OTHER STRUCTURES ........................................ III-3 A1 GROUP 1 STRUCTURES (BOILING WATER REACTOR BUILDING, PRESSURIZED WATER REACTOR SHIELD BUILDING, CONTROL ROOM/BUILDING) ................................ III A1-1 A2 GROUP 2 STRUCTURES (BOILING WATER REACTOR BUILDING WITH STEEL SUPERSTRUCTURE) ..................................... III A2-1 A3 GROUP 3 STRUCTURES (AUXILIARY BUILDING, DIESEL

GENERATOR BUILDING, RADWASTE BUILDING, TURBINE BUILDING, SWITCHGEAR ROOM, YARD STRUCTURES, SUCH AS AUXILIARY FEEDWATER PUMPHOUSE, UTILITY/PIPING TUNNELS, SECURITY/LIGHTING POLES, MANHOLES, DUCT BANKS; STATION BLACKOUT STRUCTURES, SUCH AS TRANSMISSION TOWERS, STARTUP TOWERS CIRCUIT BREAKER FOUNDATION, ELECTRICAL ENCLOSURE) ....................... III A3-1

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A4 GROUP 4 STRUCTURES (CONTAINMENT INTERNAL STRUCTURES, EXCLUDING REFUELING CANAL) .............................. III A4-1 A5 GROUP 5 STRUCTURES (FUEL STORAGE FACILITY, REFUELING CANAL) ............................................................................... III A5-1 A6 GROUP 6 STRUCTURES (WATER-CONTROL STRUCTURES) ........... III A6-1 A7 GROUP 7 STRUCTURES (CONCRETE TANKS AND MISSILE BARRIERS) ............................................................................... III A7-1 A8 GROUP 8 STRUCTURES (STEEL TANKS AND MISSILE BARRIERS) ............................................................................... III A8-1 A9 GROUP 9 STRUCTURES (BOILING WATER REACTOR UNIT VENT STACK) .......................................................................................... III A9-1 III COMPONENT SUPPORTS ....................................................................... III B-1 B1 SUPPORTS FOR ASME PIPING AND COMPONENTS ......................... III B1-1 B2 SUPPORTS FOR CABLE TRAYS, CONDUIT, HVAC DUCTS,

TUBETRACK®, INSTRUMENT TUBING, NON-ASME PIPING AND COMPONENTS ............................................................................... III B2-1

B3 ANCHORAGE OF RACKS, PANELS, CABINETS, AND ENCLOSURES FOR ELECTRICAL EQUIPMENT AND INSTRUMENTATION ............................................................................... III B3-1

B4 SUPPORTS FOR EMERGENCY DIESEL GENERATOR, HEATING VENTILATION, AND AIR CONDITIONING SYSTEM COMPONENTS, AND OTHER MISCELLANEOUS MECHANICAL EQUIPMENT ............. III B4-1

B5 SUPPORTS FOR PLATFORMS, PIPE WHIP RESTRAINTS, JET IMPINGEMENT SHIELDS, MASONRY WALLS, AND OTHER MISCELLANEOUS STRUCTURES ......................................................... III B5-1

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM ............... IV-1 A1 REACTOR VESSEL (BOILING WATER REACTOR) ............................. IV A1-1 A2 REACTOR VESSEL (PRESSURIZED WATER REACTOR) ................... IV A2-1 B1 REACTOR VESSEL INTERNALS (BOILING WATER REACTOR) ......... IV B1-1 B2 REACTOR VESSEL INTERNALS (PRESSURIZED WATER

REACTOR)—WESTINGHOUSE .............................................................. IV B2-1 B3 REACTOR VESSEL INTERNALS (PRESSURIZED WATER

REACTOR)—COMBUSTION ENGINEERING ......................................... IV B3-1 B4 REACTOR VESSEL INTERNALS (PRESSURIZED WATER

REACTOR)—BABCOCK AND WILCOX .................................................. IV B4-1 C1 REACTOR COOLANT PRESSURE BOUNDARY

(BOILING WATER REACTOR) ................................................................ IV C1-1 C2 REACTOR COOLANT SYSTEM AND CONNECTED LINES

(PRESSURIZED WATER REACTOR) ..................................................... IV C2-1 D1 STEAM GENERATOR (RECIRCULATING) ............................................. IV D1-1 D2 STEAM GENERATOR (ONCE-THROUGH) ............................................ IV D2-1 E COMMON MISCELLANEOUS MATERIAL/ ENVIRONMENT COMBINATIONS ............................................................ IV E-1

V ENGINEERED SAFETY FEATURES ........................................................................... V-1 A CONTAINMENT SPRAY SYSTEM (PRESSURIZED WATER REACTOR) ........................................................ V A-1 B STANDBY GAS TREATMENT SYSTEM (BOILING WATER REACTOR) .................................................................. V B-1 C CONTAINMENT ISOLATION COMPONENTS ........................................... V C-1

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D1 EMERGENCY CORE COOLING SYSTEM (PRESSURIZED WATER REACTOR) ...................................................... V D1-1 D2 EMERGENCY CORE COOLING SYSTEM (BOILING WATER REACTOR) ................................................................. V D2-1 E EXTERNAL SURFACES OF COMPONENTS AND MISCELLANEOUS BOLTING ..................................................................... V E-1 F COMMON MISCELLANEOUS MATERIAL/ ENVIRONMENT COMBINATIONS ............................................................ .V F-1

VI ELECTRICAL COMPONENTS .................................................................................... VI-1 A EQUIPMENT NOT SUBJECT TO 10 CFR 50.49 ENVIRONMENTAL

QUALIFICATION REQUIREMENTS .......................................................... VI A-1 B EQUIPMENT SUBJECT TO 10 CFR 50.49 ENVIRONMENTAL

QUALIFICATION REQUIREMENTS .......................................................... VI B-1

VII AUXILIARY SYSTEMS ............................................................................................... VII-1 A1 NEW FUEL STORAGE ........................................................................... VII A1-1 A2 SPENT FUEL STORAGE ........................................................................ VII A2-1 A3 SPENT FUEL POOL COOLING AND CLEANUP

(PRESSURIZED WATER REACTOR) .................................................... VII A3-1 A4 SPENT FUEL POOL COOLING AND CLEANUP

(BOILING WATER REACTOR) ............................................................... VII A4-1 A5 SUPPRESSION POOL CLEANUP SYSTEM

(BOILING WATER REACTOR) ............................................................... VII A5-1 B OVERHEAD HEAVY LOAD AND LIGHT LOAD

(RELATED TO REFUELING) HANDLING SYSTEMS ................................ VII B-1 C1 OPEN-CYCLE COOLING WATER SYSTEM

(SERVICE WATER SYSTEM) ................................................................. VII C1-1 C2 CLOSED-CYCLE COOLING WATER SYSTEM ..................................... VII C2-1 C3 ULTIMATE HEAT SINK .......................................................................... VII C3-1 D COMPRESSED AIR SYSTEM .................................................................. VII D-1 E1 CHEMICAL AND VOLUME CONTROL SYSTEM

(PRESSURIZED WATER REACTOR) .................................................... VII E1-1 E2 STANDBY LIQUID CONTROL SYSTEM

(BOILING WATER REACTOR) .............................................................. VII E2-1 E3 REACTOR WATER CLEANUP SYSTEM

(BOILING WATER REACTOR) .............................................................. VII E3-1 E4 SHUTDOWN COOLING SYSTEM

(OLDER BOILING WATER REACTOR) .................................................. VII E4-1 E5 WASTE WATER SYSTEMS .................................................................... VII E5-1 F1 CONTROL ROOM AREA VENTILATION SYSTEM ............................... VII F1-1 F2 AUXILIARY AND RADWASTE AREA VENTILATION SYSTEM ............ VII F2-1 F3 PRIMARY CONTAINMENT HEATING AND VENTILATION SYSTEM ... VII F3-1 F4 DIESEL GENERATOR BUILDING VENTILATION SYSTEM ................. VII F4-1 G FIRE PROTECTION ................................................................................. VII G-1 H1 DIESEL FUEL OIL SYSTEM ................................................................... VII H1-1 H2 EMERGENCY DIESEL GENERATOR SYSTEM ................................... VII H2-1 I EXTERNAL SURFACES OF COMPONENTS AND

MISCELLANEOUS BOLTING .................................................................... VII I-1 J COMMON MISCELLANEOUS MATERIAL/ ENVIRONMENT COMBINATIONS ............................................................ VII J-1

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VIII STEAM AND POWER CONVERSION SYSTEM .......................................................... VIII-1 A STEAM TURBINE SYSTEM ........................................................................ VIII A-1 B1 MAIN STEAM SYSTEM (PRESSURIZED WATER REACTOR) .............. VIII B1-1 B2 MAIN STEAM SYSTEM (BOILING WATER REACTOR) .......................... VIII B2-1 C EXTRACTION STEAM SYSTEM ................................................................. VIII C-1 D1 FEEDWATER SYSTEM (PRESSURIZED WATER REACTOR) .............. VIII D1-1 D2 FEEDWATER SYSTEM (BOILING WATER REACTOR) ......................... VIII D2-1 E CONDENSATE SYSTEM ............................................................................. VIII E-1 F STEAM GENERATOR BLOWDOWN SYSTEM

(PRESSURIZED WATER REACTOR) ........................................................ VIII F-1 G AUXILIARY FEEDWATER SYSTEM

(PRESSURIZED WATER REACTOR) ........................................................ VIII G-1 H EXTERNAL SURFACES OF COMPONENTS AND

MISCELLANEOUS BOLTING ...................................................................... VIII H-1 I COMMON MISCELLANEOUS MATERIAL/

ENVIRONMENT COMBINATIONS ............................................................... VIII I-1

IX USE OF TERMS FOR STRUCTURES, COMPONENTS, MATERIALS, ENVIRONMENTS, AGING EFFECTS, AND AGING MECHANISMS ............................ IX-1 A INTRODUCTION ........................................................................................... IX A-1 B STRUCTURES AND COMPONENTS ........................................................... IX B-1 C MATERIALS .................................................................................................. IX C-1 D ENVIRONMENTS ......................................................................................... IX D-1 E AGING EFFECTS ......................................................................................... IX E-1 F SIGNIFICANT AGING MECHANISMS .......................................................... IX F-1 G REFERENCES .............................................................................................. IX G-1

X AGING MANAGEMENT PROGRAMS THAT MAY BE USED TO DEMONSTRATE ACCEPTABILITY OF TIME-LIMITED AGING ANALYSES IN ACCORDANCE WITH 10 CFR 54.21(c)(1)(iii) ........................................................ X-1

X.M1 FATIGUE MONITORING ....................................................................... X.M1-1 X.M2 NEUTRON FLUENCE MONITORING ................................................... X.M2-1 X.S1 CONCRETE CONTAINMENT UNBONDED TENDON PRESTRESS .......................................................................... X.S1-1 X.E1 ENVIRONMENTAL QUALIFICATION OF ELECTRIC EQUIPMENT ....................................................................... X.E1-1

XI AGING MANAGEMENT PROGRAMS ............................................................................ XI-1 XI.M1 ASME SECTION XI INSERVICE INSPECTION, SUBSECTIONS IWB, IWC, AND IWD ................................................................................ XI.M1-1 XI.M2 WATER CHEMISTRY .............................................................................. XI.M2-1 XI.M3 REACTOR HEAD CLOSURE STUD BOLTING ...................................... XI.M3-1 XI.M4 BWR VESSEL ID ATTACHMENT WELDS ............................................. XI.M4-1 XI.M5 DELETED ................................................................................................. XI.M5-1 XI.M6 DELETED ................................................................................................. XI.M6-1 XI.M7 BWR STRESS CORROSION CRACKING ........................................... XI.M7-1 XI.M8 BWR PENETRATIONS ........................................................................ XI.M8-1 XI.M9 BWR VESSEL INTERNALS ................................................................. XI.M9-1

XI.M10 BORIC ACID CORROSION ................................................................... XI.M10-1

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XI.M11B CRACKING OF NICKEL-ALLOY COMPONENTS AND LOSS OF MATERIAL DUE TO BORIC ACID-INDUCED CORROSION IN REACTOR COOLANT PRESSURE BOUNDARY COMPONENTS (PWRs ONLY) ........................................................... XI.M11B-1

XI.M12 THERMAL AGING EMBRITTLEMENT OF CAST AUSTENITIC STAINLESS STEEL (CASS) .................................................................. XI.M12-1

XI.M16A PWR VESSEL INTERNALS ................................................................. XI.M16A-1 XI.M17 FLOW-ACCELERATED CORROSION ................................................. XI.M17-1 XI.M18 BOLTING INTEGRITY ............................................................................ XI.M18-1 XI.M19 STEAM GENERATORS ........................................................................ XI.M19-1 XI.M20 OPEN-CYCLE COOLING WATER SYSTEM ........................................ XI.M20-1 XI.M21A CLOSED TREATED WATER SYSTEMS ............................................ XI.M21A-1 XI.M22 BORAFLEX MONITORING .................................................................... XI.M22-1 XI.M23 INSPECTION OF OVERHEAD HEAVY LOAD AND LIGHT LOAD

(RELATED TO REFUELING) HANDLING SYSTEMS .......................... XI.M23-1 XI.M24 COMPRESSED AIR MONITORING ...................................................... XI.M24-1

XI.M25 BWR REACTOR WATER CLEANUP SYSTEM .................................... XI.M25-1 XI.M26 FIRE PROTECTION ............................................................................... XI.M26-1 XI.M27 FIRE WATER SYSTEM ......................................................................... XI.M27-1 XI.M29 OUTDOOR AND LARGE ATMOSPHERIC METALLIC STORAGE TANKS ................................................................................. XI.M29-1 XI.M30 FUEL OIL CHEMISTRY ......................................................................... XI.M30-1 XI.M31 REACTOR VESSEL MATERIAL SURVEILLANCE .............................. XI.M31-1 XI.M32 ONE-TIME INSPECTION ....................................................................... XI.M32-1 XI.M33 SELECTIVE LEACHING ........................................................................ XI.M33-1 XI.M35 ASME CODE CLASS 1 SMALL-BORE PIPING .................................... XI.M35-1 XI.M36 EXTERNAL SURFACES MONITORING OF MECHANICAL COMPONENTS ............................................................. XI.M36-1 XI.M37 FLUX THIMBLE TUBE INSPECTION ................................................... XI.M37-1 XI.M38 INSPECTION OF INTERNAL SURFACES IN MISCELLANEOUS PIPING AND DUCTING COMPONENTS .............................................. XI.M38-1 XI.M39 LUBRICATING OIL ANALYSIS .............................................................. XI.M39-1 XI.M40 MONITORING OF NEUTRON-ABSORBING MATERIALS OTHER THAN BORAFLEX .................................................................... XI.M40-1 XI.M41 BURIED AND UNDERGROUND PIPING AND TANKS........................ XI.M41-1 XI.M42 INTERNAL COATINGS/LININGS FOR IN-SCOPE PIPING, PIPING COMPONENTS, HEAT EXCHANGERS, AND TANKS ........... XI.M42-1 XI.S1 ASME SECTION XI, SUBSECTION IWE ................................................. XI.S1-1 XI.S2 ASME SECTION XI, SUBSECTION IWL .................................................. XI.S2-1 XI.S3 ASME SECTION XI, SUBSECTION IWF ................................................. XI.S3-1 XI.S4 10 CFR PART 50, APPENDIX J ............................................................... XI.S4-1 XI.S5 MASONRY WALLS ................................................................................... XI.S5-1 XI.S6 STRUCTURES MONITORING ................................................................ XI.S6-1 XI.S7 INSPECTION OF WATER-CONTROL STRUCTURES ASSOCIATED WITH NUCLEAR POWER PLANTS ............................... XI.S7-1 XI.S8 PROTECTIVE COATING MONITORING AND MAINTENANCE ........... XI.S8-1 XI.E1 ELECTRICAL INSULATION FOR ELECTRICAL CABLES AND CONNECTIONS NOT SUBJECT TO 10 CFR 50.49 ENVIRONMENTAL QUALIFICATION REQUIREMENTS ........................ XI.E1-1

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XI.E2 ELECTRICAL INSULATION FOR ELECTRICAL CABLES AND CONNECTIONS NOT SUBJECT TO 10 CFR 50.49 ENVIRONMENTAL QUALIFICATION REQUIREMENTS USED IN INSTRUMENTATION CIRCUITS .............................................................. XI.E2-1 XI.E3A ELECTRICAL INSULATION FOR INACCESSIBLE MEDIUM-VOLTAGE POWER CABLES NOT SUBJECT TO 10 CFR 50.49 ENVIRONMENTAL QUALIFICATION REQUIREMENTS ................................................................................... XI.E3A-1 XI.E3B ELECTRICAL INSULATION FOR INACCESSIBLE INSTRUMENT AND CONTROL CABLES NOT SUBJECT TO 10 CFR 50.49 ENVIRONMENTAL QUALIFICATION REQUIREMENTS ................................................................................... XI.E3B-1 XI.E3C ELECTRICAL INSULATION FOR INACCESSIBLE LOW-VOLTAGE POWER CABLES NOT SUBJECT TO 10 CFR 50.49 ENVIRONMENTAL QUALIFICATION REQUIREMENTS ................................................................................... XI.E3C-1 XI.E4 METAL-ENCLOSED BUS ......................................................................... XI.E4-1 XI.E5 FUSE HOLDERS ....................................................................................... XI.E5-1 XI.E6 ELECTRICAL CABLE CONNECTIONS NOT SUBJECT TO 10 CFR 50.49 ENVIRONMENTAL QUALIFICATION REQUIREMENTS ........................................................ XI.E6-1 XI.E7 HIGH-VOLTAGE INSULATORS ............................................................... XI.E7-1

APPENDIX A—QUALITY ASSURANCE FOR AGING MANAGEMENT PROGRAMS .......... A–1

APPENDIX B—OPERATING EXPERIENCE FOR AGING MANAGEMENT PROGRAMS .................................................................................................. B–1

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LIST OF TABLES

Table Page

OVERVIEW OF THE GENERIC AGING LESSONS LEARNED FOR SUBSEQUENT LICENSE RENEWAL (GALL-SLR) REPORT EVALUATION PROCESS 1 Aging Management Review Column Heading Descriptions ................................ xxxii 2 Aging Management Programs Element Descriptions ......................................... xxxiv

APPLICATION OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILER AND PRESSURE VESSEL CODE I-1 ASME Code Section XI Editions and Addenda that Are Acceptable for

Use in AMPs ............................................................................................................ I-3

CHAPTER II CONTAINMENT STRUCTURES A1 Concrete Containments (Reinforced and Prestressed) ................................... II A1-2 A2 Steel Containments .......................................................................................... II A2-2A3 Common Components ...................................................................................... II A3-2 B1.1 Mark I Steel Containments ............................................................................... II B1-2 B1.2 Mark I Concrete Containments ......................................................................... II B1-4 B2.1 Mark II Steel Containments .............................................................................. II B2-2 B2.2 Mark II Concrete Containments ........................................................................ II B2-4 B3.1 Mark III Steel Containments ............................................................................. II B3-2 B3.2 Mark III Concrete Containments ....................................................................... II B3-5 B4 Common Components ...................................................................................... II B4-2

CHAPTER III STRUCTURES AND COMPONENT SUPPORTS A1 Group 1 Structures (BWR Reactor Bldg., PWR Shield Bldg.,

Control Room/Bldg.) ........................................................................................ III A1-2 A2 Group 2 Structures (BWR Reactor Bldg. With Steel Superstructure) ............ III A2-2 A3 Group 3 Structures (Auxiliary Bldg., Diesel Generator Bldg.,

Radwaste Bldg., Turbine Bldg., Switchgear Rm., Yard Structures Such As AFW Pumphouse Utility/Piping Tunnels, Security/Lighting Poles, Manholes, Duct Banks; SBO Structures Such As Transmission Towers, Startup Tower Circuit Breaker Foundation, Electrical Enclosure) ........................................................................................ III A3-2

A4 Group Structures (Containment Internal Structures, Excluding Refueling Canal) .............................................................................................. III A4-2

A5 Group 5 Structures, (Fuel Storage Facility, Refueling Canal) ......................... III A5-2 A6 Group 6 Structures (Water-Control Structures) ............................................... III A6-2 A7 Group 7 Structures (Concrete Tanks and Missile Barriers) ............................ III A7-2 A8 Group 8 Structures (Steel Tanks and Missile Barriers) ................................... III A8-2 A9 Group 9 Structures (BWR Unit Vent Stack) .................................................... III A9-2 B1.1 Class 1 ............................................................................................................. III B1-2 B1.2 Class 2 and 3 ................................................................................................... III B1-6 B1.3 Class MC ........................................................................................................ III B1-10 B2 Support for Cable Trays, Conduit, HVAC Ducts, Tube Track,

Instrument Tubing, Non-ASME Piping and Components ............................... III B2-2 B3 Anchorage of Racks, Panels, Cabinets, and Enclosures for Electrical

Equipment and Instrumentation....................................................................... III B3-2

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B4 Supports for Emergency Diesel Generator, HVAC System Components, and Other Miscellaneous Mechanical Equipment .................... III B4-2

B5 Supports for Platforms, Pipe Whip Restraints, Jet Impingement Shields, Masonry Walls, and Other Miscellaneous Structures ....................... III B5-2

CHAPTER IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM A1 Reactor Vessel (BWR) ..................................................................................... IV A1-2 A2 Reactor Vessel (PWR) ..................................................................................... IV A2-2 B1 Reactor Vessel Internals (BWR) ...................................................................... IV B1-2 B2 Reactor Vessel Internals (PWR)—Westinghouse ........................................... IV B2-2 B3 Reactor Vessel Internals (PWR)—Combustion Engineering .......................... IV B3-2 B4 Reactor Vessel Internals (PWR)—Babcock & Wilcox .................................... IV B4-2 C1 Reactor Coolant Pressure Boundary (BWR)................................................... IV C1-2 C2 Reactor Coolant System and Connected Lines (PWR) .................................. IV C2-2 D1 Steam Generator (Recirculation) ..................................................................... IV D1-2 D2 Steam Generator (Once-Through) .................................................................. IV D2-2 E Common Miscellaneous Material/Environment Combinations ......................... IV E-2 CHAPTER V ENGINEERED SAFETY FEATURES A Containment Spray System (PWR) .................................................................... V A-2 B Standby Gas Treatment System (BWR) ............................................................ V B-2 C Containment Isolation Components ................................................................... V C-2 D1 Emergency Core Cooling System (PWR) ........................................................ V D1-2 D2 Emergency Core Cooling System (BWR) ........................................................ V D2-2 E External Surfaces of Components and Miscellaneous Bolting .......................... V E-2 F Common Miscellaneous Material/Environment Combinations .......................... V F-2 CHAPTER VI ELECTRICAL COMPONENTS A Equipment Not Subject to 10 CFR 50.49 Environmental Qualification Requirements ............................................................................... VI A-2 B Equipment Subject to 10 CFR 50.49 Environmental Qualification Requirements ............................................................................... VI B-2 CHAPTER VII AUXILIARY SYSTEMS A1 New Fuel Storage ........................................................................................... VII A1-2 A2 Spent Fuel Storage ......................................................................................... VII A2-2 A3 Spent Fuel Pool Cooling and Cleanup (PWR) ............................................... VII A3-2 A4 Spent Fuel Pool Cooling and Cleanup (BWR) ............................................... VII A4-2 B Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems ............................................................................................. VII B-2 C1 Open-Cycle Cooling Water System (Service Water System) ........................ VII C1-2 C2 Closed-Cycle Cooling Water System ............................................................. VII C2-2 C3 Ultimate Heat Sink .......................................................................................... VII C3-2 D Compressed Air System ................................................................................... VII D-2 E1 Chemical and Volume Control System (PWR) .............................................. VII E1-2 E2 Standby Liquid Control System (BWR) .......................................................... VII E2-2 E3 Reactor Water Cleanup System (BWR) ......................................................... VII E3-2 E4 Shutdown Cooling System (Older BWR) ....................................................... VII E4-2 E5 Waste Water Systems .................................................................................... VII E5-2 F1 Control Room Area Ventilation System .......................................................... VII F1-2

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F2 Auxiliary and Radwaste Area Ventilation System .......................................... VII F2-2 F3 Primary Containment Heating and Ventilation System .................................. VII F3-2 F4 Diesel Generator Building Ventilation System ............................................... VII F4-2 G Fire Protection ................................................................................................... VII G-3 H1 Diesel Fuel Oil System ................................................................................... VII H1-2 H2 Emergency Diesel Generator System ............................................................ VII H2-2 I External Surfaces of Components and Miscellaneous Bolting ......................... VII I-2 J Common Miscellaneous Material/Environment Combinations ......................... VII J-2 CHAPTER VIII STEAM AND POWER CONVERSION SYSTEM A Steam Turbine System .................................................................................... VIII A-2 B1 Main Steam System (PWR) ........................................................................... VIII B1-2 B2 Main Steam System (BWR) ........................................................................... VIII B2-2 C Extraction Steam System ................................................................................ VIII C-2 D1 Feedwater Systems (PWR) ........................................................................... VIII D1-2 D2 Feedwater Systems (BWR) ........................................................................... VIII D2-2 E Condensate System ........................................................................................ VIII E-2 F Steam Generator Blowdown System (PWR) .................................................. VIII F-2 G Auxiliary Feedwater System (PWR) ................................................................ VIII G-2 H External Surfaces of Components and Miscellaneous Bolting ....................... VIII H-2 I Common Miscellaneous Material/Environment Combinations ........................ VIII I-2 CHAPTER IX USE OF TERMS FOR STRUCTURES, COMPONENTS, MATERIALS,

ENVIRONMENTS, AGING EFFECTS, AND AGING MECHANISMS IX.B Use of Terms for Structures and Components ................................................. IX B-2 IX.C Use of Terms for Materials ................................................................................ IX C-2 IX.D Use of Terms for Environments ......................................................................... IX D-2 IX.E Use of Terms for Aging Effects .......................................................................... IX E-2 IX.F Use of Terms for Aging Mechanisms ................................................................ IX F-2 CHAPTER X AGING MANAGEMENT PROGRAMS THAT MAY BE USED TO

DEMONSTRATE ACCEPTABILITY OF TIME-LIMITED AGING ANALYSES IN ACCORDANCE WITH 10 CFR 54.21(c)(1)(III)

X-01 FSAR Supplement Summaries for GALL-SLR Report Chapter X Aging Management Programs That May Be Used to Demonstrate Acceptability of Time-Limited Aging Analyses in Accordance with 10 CFR 54.21(c)(1)(iii) ...................................................................................... X 01-1

CHAPTER XI AGING MANAGEMENT PROGRAMS XI.M12-1 Thermal Embrittlement Screening Criteria ................................................... XI.M12-2 XI.M27-1 Fire Water System Inspection and Testing Recommendations ................... XI.M27-4 XI.M29-1 Tank Inspection Recommendations ............................................................. XI.M29-4 XI.M32-1 Examples of Parameters Monitored or Inspected and Aging Effect for Specific Structure or Component .................................................................. XI.M32-3 XI.M35-1 Examinations ................................................................................................ XI.M35-2 XI.M41-1 Preventive Actions for Buried and Underground Piping and Tanks ............ XI.M41-2 XI.M41-2 Inspection of Buried and Underground Piping and Tanks ........................... XI.M41-6 XI.M41-3 Cathodic Protection Acceptance Criteria .................................................... XI.M41-11 XI.M42-1 Inspection Intervals for Internal Coatings/Linings for Tanks, Piping, Piping Components, and Heat Exchangers ................................................. XI.M42-4

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XI-01 FSAR Supplement Summaries for GALL-SLR Report Chapter XI Aging Management of Applicable Systems for SLR ................................................. XI 01-1 APPENDIX A QUALITY ASSURANCE FOR AGING MANAGEMENT PROGRAMS A-01 FSAR Supplement Summary for Quality Assurance Programs for Aging

Management Programs ......................................................................................... A-2 APPENDIX B OPERATING EXPERIENCE FOR AGING MANAGEMENT PROGRAMS B-01 FSAR Supplement Summary for Operating Experience Programs for Aging

Management Programs ......................................................................................... B-4

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LIST OF CONTRIBUTORS1 Division of License Renewal, Office of Nuclear Reactor Regulation

B. Holian Division Director C. Miller Division Director J. Lubinski Division Director G. Wilson Division Director R. Caldwell Deputy Division Director M. Delligatti Deputy Division Director J. Donoghue Deputy Division Director M. Galloway Deputy Division Director J. Marshall Deputy Director and Acting Division Director B. Beasley Acting Deputy Division Director S. Weerakkody Deputy Division Director S. Bloom Branch Chief Y. Diaz Branch Chief M. Marshall Branch Chief D. Morey Branch Chief B. Pham Branch Chief B. Wittick Branch Chief A. Hiser Senior Technical Advisor B. Brady Technical Project Manager Lead W. Burton Regulatory Project Manager Lead A. Billoch Lead Project Manager H. Jones Lead Project Manager B. Litkett Lead Project Manager J. Mitchell Lead Project Manager R. Plasse Lead Project Manager B. Rogers Lead Project Manager E. Sayoc Lead Project Manager A. Wong Lead Project Manager E. Gettys Public Coordination A. Kazi Public Coordination B. Allik Mechanical Engineering A. Bufford Structural Engineering D. Brittner Project Manager C. Doutt Electrical Engineering B. Fu Mechanical Engineering W. Gardner Mechanical Engineering

1The titles in this List of Contributions refer to the NRC staff’s role in the development of this document, not their current position.

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J. Gavula Mechanical Engineering B. Grange Project Manager K. Green Mechanical Engineering W. Holston Mechanical Engineering C. Hovanec Materials Engineering R. Kalikian Mechanical Engineering J. Medoff Mechanical Engineering S. Min Materials Engineering A. Prinaris Structural Engineering M. Sadollah Electrical Engineering G. Thomas Structural Engineering M. Yoo Mechanical Engineering

Other Divisions in the Office of Nuclear Reactor Regulation D. Alley Branch Chief S. Bailey Branch Chief R. Dennig Branch Chief C. Jackson Branch Chief A. Klein Branch Chief G. Kulesa Branch Chief T. Lupold Branch Chief J. McHale Branch Chief S. Rosenberg Branch Chief J. Zimmerman Branch Chief R. Hardies Senior Level Advisor K. Karwoski Senior Level Advisor L. Banic Project Manager G. Cheruvenki Materials Engineering J. Collins Materials Engineering S. Cumblidge Materials Engineering A. Erickson Structural Engineering C. Fairbanks Materials Engineering M. Hardgrove Mechanical Engineering K. Hoffman Materials Engineering N. Iqbal Fire Protection Engineering A. Johnson Reactor Operations Engineering S. Jones Reactor Systems Engineering B. Lee Reactor Systems Engineering B. Lehman Structural Engineering R. Mathew Electrical Engineering C. Ng Mechanical Engineering D. Nguyen Electrical Engineering

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A. Obodoako Materials Engineering A. Patel Reactor Engineering B. Parks Reactor Engineering J. Poehler Materials Engineering P. Purtscher Materials Engineering S. Ray Electrical Engineering S. Sheng Materials Engineering A. Tsirigotis Mechanical Engineer P. Verdi International Assignee O. Yee Reactor Systems Engineering M. Yoder Chemical Engineering

Region II P. Cooper Sr. Reactor Inspector J. Rivera-Ortiz Reactor Inspector

Region III N. Feliz-Adorno Sr. Reactor Inspector M. Holmberg Sr. Reactor Inspector C. Tilton Sr. Reactor Inspector

Region IV S. Graves Sr. Reactor Inspector G. Pick Sr. Reactor Inspector M. Williams Reactor Inspector

Office of New Reactors J. Xu Branch Chief A. Istar Structural Engineering

Office of Nuclear Material Safety and Safeguards A. Csontos Branch Chief J. Wise Materials Engineering

Office of Nuclear Regulatory Research J. Burke Branch Chief S. Frankl Branch Chief M. Gavrilas Branch Chief J. Nakoski Branch Chief W. Ott Branch Chief D. Rudland Branch Chief M. Salley Branch Chief R. Sydnor Branch Chief J. Ake Senior Technical Advisor—Geophysical Engineering T. Nicholson Senior Technical Advisor—Radionuclide Transport R. Tregoning Senior Technical Advisor—Materials Engineering A. Hull Team Leader

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K. Arai Materials Engineering M. Benson Materials Engineering E. Focht Materials Engineering M. Fuhrman Geochemistry C. Harris Materials Engineering M. Hiser Materials Engineering M. Homiack Mechanical Engineering M. Kirk Materials Engineering B. Lin Mechanical Engineering S. Malik Materials Engineering K. Miller Electrical Engineering W. Norris Materials Engineering G. Oberson Materials Engineering R. Perkins Reliability & Risk Engineering I. Prokofiev Materials Engineering J. Philip Geotechnical Engineering A. Pulvirenti Materials Engineering S. Rao Materials Engineering M. Rossi Materials Engineering M. Sircar Structural Engineering M. Srinivasan Materials Engineering G. Stevens Materials Engineering D. Stroup Fire Protection Engineering G. Wang Mechanical Engineering

Center for Nuclear Waste Regulatory Analyses, Southwest Research Institute® G. Adams Computer/Industrial Engineering L. Howard Project Manager/Nuclear Engineering L. Naukam Program Support/Technical Editing Y. Pan Materials Engineering A. Ramos Program Support/Technical Editing D. Speaker Nuclear Engineering

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ABBREVIATIONS ACAR aluminum conductor aluminum alloy reinforced ACSR aluminum conductor steel reinforced ACI American Concrete Institute ADAMS Agencywide Documents Access and Management System AEA Atomic Energy Act AEC Atomic Energy Commission AFW auxiliary feedwater AERM aging effect requiring management AISC American Institute of Steel Construction Al Aluminum AMPs aging management programs AMR aging management review ANSI American National Standards Institute API American Petroleum Institute ASCE American Society of Civil Engineers ASME American Society of Mechanical Engineers ASME Code American Society of Mechanical Engineers Boiler and Pressure Vessel Code ASTM ASTM International AWG American wire gauge B&W Babcock & Wilcox BWR boiling water reactor BWRVIP Boiling Water Reactor Vessel and Internals Project CASS cast austenitic stainless steel CB core barrel CCCW closed-cycle cooling water CE Combustion Engineering CEA control element assembly CFR Code of Federal Regulations CFS core flood system CLB current licensing basis CRD control rod drive CRGT control rod guide tube CSE copper/copper sulfate reference electrode CVCS chemical and volume control system DOE U.S. Department of Energy DSCSS drywell and suppression chamber spray system

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ECT eddy current testing EDG emergency diesel generator EMDA Expanded Materials Degradation Assessment EPDM ethylene propylene diene monomer EPR ethylene-propylene rubber EPRI Electric Power Research Institute EQ environmental qualification FAC flow-accelerated corrosion FERC Federal Energy Regulatory Commission FRN Federal Register Notice FSAR Final Safety Analysis Report FW feedwater GALL Generic Aging Lessons Learned GALL-SLR Generic Aging Lessons Learned for Subsequent License Renewal GL generic letter HDPE high density polyethylene HELB high-energy line break HP high pressure HPCI high-pressure coolant injection HPCS high-pressure core spray HPSI high-pressure safety injection HVAC heating, ventilation, and air conditioning I&C instrumentation and control I&E inspection and evaluation IAEA International Atomic Energy Agency IASCC irradiation-assisted stress corrosion cracking IC isolation condenser ID inside diameter IEB Inspection and Enforcement Bulletin IEEE Institute of Electrical and Electronics Engineers IGA intergranular attack IGSCC intergranular stress corrosion cracking IMI incore monitoring instrumentation IN information notice INPO Institute of Nuclear Power Operations IRM intermediate range monitor

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ISA International Society of Automation ISG interim staff guidance ISI inservice inspection ISP integrated surveillance program LERs licensee event reports LG lower grid LOCA loss of coolant accident LP low pressure LPCI low-pressure coolant injection LPCS low-pressure core spray LPSI low-pressure safety injection LR-ISG license renewal interim staff guidance LRT leak rate test LWR light water reactor MIC microbiologically influenced corrosion MPa megapascal MRP Materials Reliability Program MS main steam MSR moisture separator/reheater NACE National Association of Corrosion Engineers NDE nondestructive examination NEA Nuclear Energy Agency NEI Nuclear Energy Institute NFPA National Fire Protection Association NPP nuclear power plant NPS nominal pipe size NRC U.S. Nuclear Regulatory Commission NSAC Nuclear Safety Analysis Center NUMARC Nuclear Management and Resources Council OCCW open-cycle cooling water ODSCC outside diameter stress corrosion cracking OECD Organisation for Economic Co-operation and Development OE operating experience PVC polyvinyl chloride PWR pressurized water reactor PWSCC primary water stress corrosion cracking

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QA quality assurance RCIC reactor core isolation cooling RCP reactor coolant pump RCPB reactor coolant pressure boundary RCS reactor coolant system RES Office of Nuclear Regulatory Research RG Regulatory Guide RHR residual heat removal RVI reactor vessel internal RWCU reactor water cleanup RWT refueling water tank SBO SCs

station blackout structures and components

SCC stress corrosion cracking SDC shutdown cooling SEI Structural Engineering Institute SFP spent fuel pool SG steam generator S/G standards and guides SIT safety injection tank SLC standby liquid control SLR subsequent license renewal SLRAs subsequent license renewal applications SOC Statements of Consideration SOER significant operating experience report SRM source range monitor SRM staff requirements memorandum SRP-SLR Standard Review Plan for Review of Subsequent License Renewal Applications for

Nuclear Power Plants SS stainless steel SSCs systems, structures, and components

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TGSCC transgranular stress corrosion cracking TLAA time-limited aging analysis TS technical specifications UHS ultimate heat sink USACE U.S. Army Corps of Engineers USE upper-shelf energy UT ultrasonic testing UV ultraviolet XLPE cross-linked polyethylene

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INTRODUCTION

NUREG–2191, “Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report,” is referenced as a technical basis document in NUREG–2192, “Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants,” (SRP-SLR). The GALL-SLR Report lists generic aging management reviews of systems, structures, and components (SSCs) that may be in the scope of subsequent license renewal applications (SLRAs) and identifies aging management programs (AMPs) that are determined to be acceptable to manage aging effects of SSCs in the scope of license renewal, as required by Title 10 of the Code of Federal Regulations (10 CFR) Part 54, “Requirements for Renewal of Operating Licenses for Nuclear Power Plants.” If an applicant credits an AMP described in the GALL-SLR Report in the SLRA, the applicant should ensure that the conditions and operating experience (OE) at the plant are bounded by the conditions and OE for which the GALL-SLR Report program was evaluated. If these bounding conditions are not met, the applicant should address any additional aging effects and augment the AMPs for subsequent license renewal. If an SLRA references the GALL-SLR Report as the approach used to manage aging effect(s), the U.S. Nuclear Regulatory Commission staff will use the GALL-SLR Report as a basis for the SLRA assessment consistent with guidance specified in the SRP-SLR.

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BACKGROUND

The Atomic Energy Act (AEA) of 1954, as amended, allows the U.S. Nuclear Regulatory Commission (NRC) to issue licenses for commercial nuclear power reactors to operate for up to 40 years. The NRC regulations permit these licenses to be renewed beyond the initial 40-year term for an additional period of time, limited to 20-year increments per renewal, based on the results of an assessment to determine if the nuclear facility can continue to operate safely during the proposed period of extended operation. There are no limitations in the AEA or the NRC regulations restricting the number of times a license may be renewed.

The focus of license renewal, as described in Title 10 of the Code of Federal Regulations (10 CFR) Part 54, is to identify aging effects that could impair the ability of systems, structures, and components (SSCs) within the scope of license renewal to perform their intended functions, and to demonstrate that these effects will be adequately managed during the period of extended operation. The regulatory requirements for both initial and subsequent license renewal (SLR) are established by 10 CFR Part 54. To address the unique aspects of material aging and degradation that would apply to SLR (e.g., to permit plants to operate to 80 years), the Office of Nuclear Reactor Regulation requested support from the Office of Nuclear Regulatory Research (RES) to develop technical information to evaluate the feasibility of SLR. RES has memoranda of understanding with both the U.S. Department of Energy (DOE) and the Electric Power Research Institute to cooperate in nuclear safety research related to long-term operations beyond 60 years. Under these memoranda, the NRC and the DOE held two international conferences, in 2008 and 2011, on reactor operations beyond 60 years. In May 2012, the NRC and the DOE also co-sponsored the Third International Conference on Nuclear Power Plant Life Management for Long-Term Operations, organized by the International Atomic Energy Agency (IAEA). In February 2013 and February 2015, the Nuclear Energy Institute (NEI) held a forum on long-term operations and SLR. These conferences laid out the technical issues that would need to be addressed to provide assurance for safe operation beyond 60 years.

Based on the information gathered from these conferences and forums, and from other sources over the past several years, the most significant technical issues identified as challenging operation beyond 60 years are: reactor pressure vessel embrittlement; irradiation-assisted stress corrosion cracking (SCC) of reactor internals; concrete structures and containment degradation; and electrical cable environmental qualification, condition monitoring and assessment. Throughout this process, the NRC staff has emphasized that it is the industry’s responsibility to resolve these and other issues to provide the technical bases to ensure safe operation beyond 60 years.

The NRC, in cooperation with the DOE, completed the Expanded Materials Degradation Assessment (EMDA) in 2014 [Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML14279A321, ML14279A331, ML14279A349, ML14279A430, and ML14279A461]. The EMDA uses an expert elicitation process to identify materials and components, which could be susceptible to significant degradation during operation beyond 60 years. The EMDA covers the reactor vessel, primary system piping, reactor vessel internals, concrete, and electrical cables and qualification. The NRC staff used the results of the EMDA to identify gaps in the current technical knowledge or issues not being addressed by planned industry or DOE research, and to identify aging management programs (AMPs) that will require modification for SLR.

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On May 9, 2012 (ADAMS Accession No. ML12158A545) and subsequently on November 1, 13, and 14, 2012, the NRC staff and interested stakeholders met to discuss issues and receive comments for consideration for SLR. The staff’s resolution to these public comments is available in the staff’s memo dated September 12, 2016 (ADAMS Accession No. ML16194A222).

In addition to working with external stakeholders, the NRC staff conducted AMP effectiveness audits at three units that were at least 2 years into the period of extended operation. The purpose of these information gathering audits was to better understand how licensees are implementing the license renewal AMPs, in terms of both the findings and the effectiveness of the programs, and to develop recommendations for updating license renewal guidance. The NRC staff used the information gathered from these audits to update the SLR guidance based on the staff’s experience with the aging management activities during the first license renewals. A summary of the first two AMP effectiveness audits can be found in the May 2013 report, “Summary of Aging Management Program Effectiveness Audits to Inform Subsequent License Renewal: R.E. Ginna NPP and Nine Mile Point Nuclear Station, Unit 1” (ADAMS Accession No. ML13122A007). The summary of the third audit can be found in the August 5, 2014, report, “H.B. Robinson Steam Electric Plant, Unit 2, Aging Management Program Effectiveness Audit” (ADAMS Accession No. ML14017A289). In addition, on June 15, 2016, the staff issued the Technical Letter Report, “Review of Aging Management Programs: Compendium of Insight from License Renewal Applications and from AMP Effectiveness Audits Conducted to Inform Subsequent License Renewal Guidance Documents,” (ADAMS Accession No. ML16167A076), which provides the staff’s observations from reviewing license renewal applications and the AMP effectiveness audits.

The NRC staff reviewed domestic operating experience (OE) as reported in licensee event reports and NRC generic communications related to failures and degradation of passive components. Similarly the NRC staff reviewed the following international OE databases: (i) International Reporting System, jointly operated by the IAEA; (ii) IAEA’s International Generic Ageing Lessons Learned Programme; (iii) Organisation for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) Component Operational Experience and Degradation and Ageing Programme database; and (iv) OECD/NEA Cable Aging Data and Knowledge database.

The NRC staff reviewed the results from AMP audits, findings from the EMDA, domestic and international OE, and public comments to identify technical issues that need to be considered for assuring the safe operation of NRC-licensed nuclear power plants (NPPs). By letter dated August 6, 2014 (ADAMS Accession No. ML14253A104), NEI documented the industry’s views and recommendations for updating NUREG–1801, Revision 2, “Generic Aging Lessons Learned (GALL) Report,” and NUREG–1800, Revision 2, “Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants,” to support SLR. Between fiscal years 2014 and 2015, the NRC staff reviewed the comments and recommendations and drafted the Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report to ensure that sufficient guidance was in place to support review of an SLR application in 2018 or 2019.

The staff requirements memorandum (SRM) on SECY-14-0016, “Ongoing Staff Activities to Assess Regulatory Considerations for Power Reactor Subsequent License Renewal,” (ADAMS Accession No. ML14241A578) directed the staff to continue to update the license renewal guidance, as needed, to provide additional clarity on the implementation of the license renewal regulatory framework. The SRM also directed the staff to keep the Commission informed on the progress in resolving the following technical issues related to SLR: (i) reactor pressure vessel neutron embrittlement at high fluence, (ii) irradiation-assisted SCC of reactor internals

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and primary system components, (iii) concrete and containment degradation, and (iv) electrical cable qualification and condition assessment. In addition, the SRM directed that the staff should keep the Commission informed regarding the staff’s readiness for accepting an application and any further need for regulatory process changes, rulemaking, or research.

During the staff’s consideration of revisions to 10 CFR Part 54, changes were considered to the License Renewal Rule to address the provisions of 10 CFR 50.54(hh)(2) regarding guidance and strategies to maintain and restore core cooling, containment, and spent fuel cooling capabilities under the circumstances associated with the loss of large areas of the plant due to explosions or fires. After discussions with stakeholders and the public, it was concluded that these issues need not be addressed in the License Renewal Rule because emergency preparedness equipment is not identified in 10 CFR 54.4(a)(3). The 1995 Federal Register Notice for the final license renewal rule, 60 FR 22461, 22468 states:

Regarding systems, structures, and components required to make protective action recommendations, the Commission thoroughly evaluated emergency planning considerations in the previous license renewal rulemaking. These evaluations and conclusions are still valid and can be found in the [Statements of Consideration] SOC for the previous license renewal rule (56 FR 64943 at 64966). Therefore, the Commission concludes that systems, structures, and components required for emergency planning, unless they meet the scoping criteria in §54.4, should not be the focus of a license renewal review.

Further, even if this equipment is within the scope of license renewal that does not necessarily mean that it is subject to aging management review based on the existing rule in that only passive, long-lived structures and components are subject to an aging management review. Further, this is not an issue specific to SLR and is inconsistent with the first principle of license renewal (i.e., “….with the exception of age-related degradation and possibly a few other issues related to safety only during extended operation of nuclear power plants, the existing regulatory process is adequate to ensure that the licensing bases of all currently operating plants provide and maintain an acceptable level of safety so that operation will not be inimical to public health and safety or common defense and security”). Therefore, there is no need to address 10 CFR 50.54(hh) and diverse and flexible mitigation capability equipment in the License Renewal Rule.

The GALL-SLR report also includes the NRC staff’s resolutions of License Renewal Interim Staff Guidance (LR-ISGs) from 2011 through 2016. Under the LR-ISG process, the NRC staff, industry, or stakeholders can propose a change to certain license renewal guidance documents. The NRC staff evaluates the issue, develops the proposed LR-ISG, issues it for public comment, evaluates any comments received, and, if necessary, issues the final LR-ISG.

The LR-ISG is then used until the NRC staff incorporates the revised guidance into a formal license renewal guidance document revision. The LR-ISGs addressed in the GALL-SLR report are:

LR-ISG-2011-01: Aging Management of Stainless Steel Structures and Components in Treated Borated Water, Revision 1

LR-ISG-2011-02: Aging Management Program for Steam Generators

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LR-ISG-2011-03: Generic Aging Lessons Learned (GALL) Report Revision 2 AMP XI.M41, “Buried and Underground Piping and Tanks”

LR-ISG-2011-04: Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water Reactors

LR-ISG-2011-05: Ongoing Review of Operating Experience

LR-ISG-2012-01: Wall Thinning Due to Erosion Mechanisms

LR-ISG-2012-02: Aging Management of Internal Surfaces, Fire Water Systems, Atmospheric Storage Tanks, and Corrosion Under Insulation

LR-ISG-2013-01: Aging Management of Loss of Coating or Lining Integrity for Internal Coatings/Linings on In-Scope Piping, Piping Components, Heat Exchangers, and Tanks

LR-ISG-2015-01: Changes to Buried and Underground Piping and Tank Recommendations

LR-ISG-2016-01: Changes to Aging Management Guidance for Various Steam Generator Components

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OVERVIEW OF THE GENERIC AGING LESSONS LEARNED FOR SUBSEQUENT LICENSE RENEWAL (GALL-SLR) REPORT

EVALUATION PROCESS

The Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report contains 11 chapters and 2 appendices. The majority of the chapters contain summary descriptions and tabulations of evaluations of aging management programs (AMPs) for a large number of structures and components (SCs) in major plant systems found in light-water reactor nuclear power plants. The major plant systems include the containment structures (Chapter II), SC supports (Chapter III), reactor vessel, internals and reactor coolant system (Chapter IV), engineered safety features (Chapter V), electrical components (Chapter VI), auxiliary systems (Chapter VII), and steam and power conversion system (Chapter VIII).

Chapter I of the GALL-SLR Report addresses the application of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code for subsequent license renewal (SLR). Chapter IX contains the description of a selection of standard terms used within the GALL-SLR Report. Chapter X contains examples of AMPs that may be used to demonstrate the acceptance of time-limited aging analyses (TLAAs) in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 54.21(c)(1)(iii). Chapter XI contains the AMPs for the mechanical, structural and electrical components. The appendices of the GALL-SLR Report address quality assurance for AMPs and operating experience (OE).

The evaluation process for the AMPs and the application of the GALL-SLR Report is described in this document. The aging management review (AMR) items for the GALL-SLR Report are presented in tabular format as described in Table 1. Table 1 describes the information presented in each column of the tables in Chapters II through VIII contained in this report.

The staff’s evaluation of the adequacy of each generic AMP to manage certain aging effects for particular SCs is based on its review of the 10 program elements in each AMP, as defined in Table 2.

On the basis of its evaluation, if the staff determines that a program is adequate to manage certain aging effects for a particular SC without change, the “Further Evaluation” entry will indicate that no further evaluation is recommended for SLR.

Chapters X and XI of the GALL-SLR Report contain generic AMPs that the staff finds to be sufficient to manage aging effects in the subsequent period of extended operation, such as the ASME Code Section XI inservice inspection, water chemistry, or structures monitoring program.

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Table 1. Aging Management Review Column Heading Descriptions Column Heading Description

New (N), Modified (M), Deleted (D), Edited (E) Item

Identifies the item as new to GALL-SLR Report, modified from GALL Revision 2, deleted from GALL Revision 2, edited from GALL Revision 2, or if blank, is unchanged from GALL Revision 2.

Item Identifies a unique number for the item (i.e., VII.G.A-91). The first part of the number indicates the chapter and AMR system (e.g., VII.G is in the auxiliary systems, fire protection system), and the second part is a unique chapter-specific identifier within a chapter (e.g., A–91 for auxiliary systems).

Standard Review Plan (SRP) Item (Table, ID)

For each row in the subsystem tables, this item identifies the corresponding row identifier from the SRP-SLR to provide the crosswalk to the SRP system table items.

Structure and/or Component Identifies the structure or components to which the row applies. Material Identifies the material of construction. See Chapter IX.C of this

report for further information. Environment Identifies the environment applicable to this row.

See Chapter IX.D of this report for further information. Aging Effect/ Mechanism Identifies the applicable aging effect and mechanism(s).

See Chapters IX.E and IX.F of this report for more information on applicable aging effects/mechanisms.

Aging Management Program (AMP)/TLAA

Identifies an AMP/TLAA found acceptable for adequately managing the effects of aging. See Chapters X and XI of this report.

Further Evaluation Identifies whether a further evaluation is needed.

Edited (E) items, in contrast to modified (M) items, are those for which no technical aspects were changed. Examples of editorial changes include:

Line item citations that were missed in the SRP SLR Table 3.X-1.

Deleting whether the environment is internal or external from the description of the environment because based on the material, environment, aging effect, and AMP combination, it is obvious that the environment could only be on either the inside or outside of the component.

Deletion of the term “piping element” from aging management review items that do not cite glass as a material. Piping elements were defined in the GALL Report as components constructed of glass.

Line item changes that only involved removing detail related to a Further Evaluation Recommended column was removed after it was verified that the identical information was included in the SRP-SLR further evaluation section.

Line item changes that only involved renumbering further evaluation sections.

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Aging effects changed from “and” to “or.” This could appear to be a technical change; however, this is not the case because the staff confirmed that is was never the intent that both aging effects were occurring. For example, the “and” in cracking due to stress corrosion cracking and cyclic loading was replaced with “or.”

Deleting the term “environment” from the description of the environment in the “Environment” column when the phrase “any environment” was used because it was obvious and redundant.

Descriptors for the AMPs in the “Aging Management Program/TLAA” column were simplified if the information was provided elsewhere.

Minor edits to component descriptions, examples: (a) deleting “elastomer” from “elastomer, elastomer seals;” (b) adding “piping” or “ducting” in front of the term “component.”

Adding the term “electrical” to Structure and/or Component and Aging Effect/Mechanism description.

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Table 2. Aging Management Programs Element Descriptions AMP Element Description

1. Scope of the Program The scope of the program should include the specific structures and components subject to an AMR.

2. Preventive Actions Preventive actions should mitigate or prevent the applicable aging effects.

3. Parameters Monitored or Inspected

This identifies the aging effects that the program manages and provides a link between the parameter or parameters that will be monitored and how the monitoring of these parameters will maintain adequate aging management.

4. Detection of Aging Effects Detection of aging effects should occur before there is a loss of any structure and component intended function. This element describes aspects such as method or technique (i.e., visual, volumetric, surface inspection), frequency, sample size, data collection, and timing of new/one-time inspections to ensure timely detection of aging effects.

5. Monitoring and Trending Monitoring and trending should provide for an estimate of the extent of the effects of aging and timely corrective or mitigative actions.

6. Acceptance Criteria Acceptance criteria, against which the need for corrective action will be evaluated, should provide reasonable assurance that the particular structure and component’s intended functions are maintained under all current licensing basis conditions during the subsequent period of extended operation.

7. Corrective Actions Description of corrective actions that will be implemented if the acceptance criteria of the program are not met.

8. Confirmation Process The confirmation process should provide reasonable assurance that preventive actions are adequate and that appropriate corrective actions have been completed and are effective.

9. Administrative Controls Administrative controls should provide a formal review and approval process.

10. Operating Experience (OE) OE applicable to the AMP, including past corrective actions resulting in program enhancements or additional programs, should provide objective evidence to support the conclusion that the effects of aging will be managed adequately so that the structure- and component intended function(s) will be maintained during the subsequent period of extended operation. In addition, an ongoing review of both plant-specific and industry OE provides reasonable assurance that the AMP is effective in managing the aging effects for which it is credited. The AMP is either enhanced or new AMPs are developed, as appropriate, when it is determined through the evaluation of OE that the effects of aging may not be adequately managed.

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EXPLANATION OF THE USE OF MULTIPLE AGING MANAGEMENT PROGRAMS IN AGING MANAGEMENT REVIEW ITEMS

For aging management review (AMR) items associated with some “Further Evaluations,” the associated Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report items now include a letter suffix with the unique chapter-specific identifier. For these items, the staff designated the various aging management programs (AMPs) it found to be acceptable in lieu of specifying “plant-specific aging management program” in the Aging Management Program column. Depending on the GALL-SLR Report Table 2 item cited in the subsequent license renewal application (SLRA) for these items, applicants can either use one of the AMPs found to be acceptable to the staff for specific situations or, comparable to any other item, can propose their own plant-specific program to manage the associated aging effect.

For example, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants (SRP-SLR) Section 3.1.2.2.16 is a further evaluation associated with SRP-SLR item 3.1-1, 136, for loss of material due to pitting and crevice corrosion in stainless steel and nickel alloy piping and piping components. The associated chapter-specific identifier has been expanded to include items R-452a, R-452b, R-452c, and R-452d. The further evaluation recommends a review of plant-specific operating experience (OE) to determine if the site’s air environments are sufficiently aggressive to cause pitting and crevice corrosion. The need to manage this aging effect will depend on the results of the OE reviews and a one-time inspection to demonstrate that pitting and crevice corrosion are not occurring or are occurring sufficiently slowly. Consequently, the acceptable AMP could be XI.M32 for performing the one-time inspection (if the aging effect does not need to be periodically managed), or it could be XI.M36, XI.M38, or XI.M42, depending on whether a periodic program is needed for external surfaces, internal surfaces, or coatings/linings. The SLRA will specify the applicable AMP by citing the specific GALL-SLR item R-452a, R-452b, R-452c, or R-452d for the corresponding AMP being used at the site. More specifically, if the plant-specific OE review does not reveal any instances of loss of material for stainless steel or nickel alloy piping and piping components, R-452a (AMP XI.M32) would be the cited SLRA AMR Table 2 item. In contrast, if external loss of material has occurred, and it was sufficient to potentially affect the intended function, R-452b (AMP XI.M36) or R-452d (AMP XI.M42) would be cited.

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REFERENCES

References are listed in the aging management program (AMP) following the program elements. References consist of documents (e.g., Codes, Standards) associated with recommended actions (e.g., qualification of personnel, inspection methods) cited in the program elements or documents containing background information associated with the AMP (e.g., Information Notices). The specific version (e.g., edition, addenda, revision) of a reference is cited in the list of references. It should be noted that in some instances, specific program elements might cite a different version of a reference than that cited in the reference list. In these cases, the staff has reviewed the provisions of the different version of the reference and has specifically cited a version based on the requirements or guidance contained within the document. Where a specific version is not cited in a program element, the version cited in the reference list is applicable. With the exception of the guidance on use of later editions/revisions of various industry documents cited below, an applicant should identify exceptions to the Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report and provide justification when using a different version of a reference cited in the program elements.

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GUIDANCE ON USE OF LATER EDITIONS/REVISIONS OF VARIOUS INDUSTRY DOCUMENTS

To aid applicants in the development of their subsequent license renewal applications (SLRAs), the staff has developed a list of aging management programs in the Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report that are based entirely or in part on specific editions/revisions of various industry codes (other than the American Society of Mechanical Engineers Boiler and Pressure Vessel Code), standards, and other industry-generated guidance documents. SLRAs may use later editions/revisions of these industry generated documents, subject to the following provisions:

(i) If the later edition/revision has been explicitly reviewed and approved/endorsed by the U.S. Nuclear Regulatory Commission (NRC) staff for license renewal via a NRC Regulatory Guide endorsement, a safety evaluation for generic use [such as for a Boiling Water Reactor Vessel and Internals Project (BWRVIP)], incorporation into Title 10 of the Code of Federal Regulation (10 CFR), or license renewal interim staff guidance.

(ii) If the later edition/revision has been explicitly reviewed and approved on a plant-specific basis by the NRC staff in its Safety Evaluation Report for another applicant’s SLRA (a precedent exists). Applicants may reference this and justify applicability to their facility via the exception process in Nuclear Energy Institute 95-10.

If either of these methods is used as justification for adopting a later edition/revision than specified in the GALL-SLR Report, the applicant shall reference the information pertaining to the NRC endorsement/approval of the later edition/revision.

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APPLICATION OF THE GENERIC AGING LESSONS LEARNED FOR SUBSEQUENT LICENSE RENEWAL (GALL-SLR) REPORT

The Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report is a technical basis document to the Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants (SRP-SLR), which provides the staff with guidance in reviewing a subsequent license renewal application (SLRA). The GALL-SLR Report should be treated in the same manner as an approved topical report that is generically applicable. An applicant may reference the GALL-SLR Report in an SLRA to demonstrate that the aging management programs (AMPs) at the applicant’s facility correspond to those reviewed and approved in the GALL-SLR Report.

If an applicant takes credit for an AMP in GALL-SLR Report, it is incumbent on the applicant to ensure that the plant AMP contains all the elements of the referenced GALL-SLR program. In addition, the conditions and operating experience (OE) at the plant must be bounded by the conditions and OE for which the GALL-SLR Report AMP was evaluated; otherwise it is incumbent on the applicant to augment the GALL-SLR Report AMP as appropriate to address the impact of the plant-specific OE on the AMP element criteria. The documentation for the above verifications must be available onsite in an auditable form.

The GALL-SLR Report contains one acceptable way to manage aging effects for subsequent license renewal (SLR). An applicant may propose alternatives for staff review in its plant-specific SLRA. The use of the GALL-SLR Report is not required, but its use should facilitate both preparation of an SLRA by an applicant and timely, consistent review by the U.S. Nuclear Regulatory Commission staff.

The GALL-SLR Report does not address scoping of structures and components (SCs) for license renewal; this is addressed in SRP-SLR Chapter 2. Scoping is plant-specific, and the results depend on the plant design and current licensing basis. The inclusion of a certain structure or component in the GALL-SLR Report does not imply that this particular structure or component is within the scope of license renewal for all plants. Conversely, the omission of a certain structure or component in the GALL-SLR Report does not imply that this particular structure or component is not within the scope of SLR for any plants.

The GALL-SLR Report contains an evaluation of a large number of SCs that may be in the scope of a typical SLRA. The evaluation results documented in the GALL-SLR Report indicate that many existing, typical generic AMPs are adequate to manage aging effects for particular structures or components for SLR without change. The GALL-SLR Report also contains recommendations on specific areas for which existing generic AMPs should be augmented (require further evaluation) for SLR and documents the technical basis for each such determination. The GALL-SLR Report identifies certain systems, structures, and components (SSCs) that may or may not be subject to particular aging effects, and those for which industry is developing generic AMPs or investigating whether aging management is warranted.

Appendix A of the GALL-SLR Report addresses quality assurance (QA) for AMPs. Those aspects of the aging management review (AMR) process that affect the quality of safety-related SSCs are subject to the QA requirements of Appendix B to Title 10 of the Code of Federal Regulations (10 CFR) Part 50. For nonsafety-related SCs subject to an AMR, the existing 10 CFR Part 50, Appendix B, QA program may be used by an applicant to address the elements of the corrective actions, confirmation process, and administrative controls for an AMP for SLR.

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The GALL-SLR Report provides a technical basis for crediting existing plant AMPs and recommending areas for AMP augmentation and further evaluation. The incorporation of the GALL-SLR Report information into the SRP-SLR, as directed by the Commission, should improve the efficiency of the SLR review process and the use of staff resources.

CHAPTER I

APPLICATION OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILER AND PRESSURE VESSEL CODE

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I APPLICATION OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILER AND PRESSURE VESSEL CODE

The American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Division 1, Sections III (design) and XI (inservice inspection requirements) were developed and are revised periodically by industry code committees composed of representatives of utilities, reactor designers, architect-engineers, component manufacturers, insurance companies, the U.S. Nuclear Regulatory Commission (NRC), and others. In 1971, the Atomic Energy Commission (AEC), the predecessor of the NRC, incorporated the ASME Code into the regulations in Title 10 of the Code of Federal Regulations (10 CFR) 50.55a through issuance of the Federal Register Notice (FRN) for the final rule (36 FR 11423 [June 12, 1971]).

The statements of consideration (SOC) for the initial issuance of 10 CFR 50.55a provides the bases for AEC’s endorsement and use of the ASME Code:

It has been generally recognized that, for boiling and pressurized water-cooled reactors, pressure vessels, piping, pumps, and valves which are part of the reactor coolant pressure boundary should, as a minimum, be designed, fabricated, inspected, and tested in accordance with the requirements of the applicable American Society of Mechanical Engineers (ASME) codes in effect at the time the equipment is purchased[.]

Because of the safety significance of uniform early compliance by the nuclear industry with the requirements of these ASME codes and published code revisions, the Commission has adopted the following amendments to Part 50 and 115, which require that certain components and systems of water-cooled reactors important to safety comply with these codes and appropriate revisions to the codes at the earliest feasible time.

Compliance with the provisions of the amendments and the referenced codes is intended to insure a basic, sound quality level.

These ASME Code sections are based on the collective engineering judgment of the code committees and document the conditions that must be monitored, the inspection techniques to identify those conditions, the frequency of the inspections, and the acceptance criteria that the inspection results must meet in order to assure the integrity of the structures and components considered in the code. The NRC has accepted this engineering judgment by endorsing the use of selected sections of the ASME Code, as incorporated in 10 CFR 50.55a.

In addition, the NRC periodically amends 10 CFR 50.55a and issues FRNs about this rule in order to endorse, by reference, newer editions and ASME Code addenda subject to the modifications and limitations identified in 10 CFR 50.55a. As stated in 65 FR 53050 (August 31, 2000):

To ensure that the GALL report conclusions will remain valid when future editions of the ASME Code are incorporated into the NRC regulations by the 10 CFR 50.55a rulemaking, the staff will perform an evaluation of these later editions for their adequacy for license renewal using the 10-element program evaluation described in the GALL Report as part of the 10 CFR 50.55a rulemaking.

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The staff will continue to evaluate future editions of the ASME Code for their adequacy for subsequent license renewal, and will document this evaluation in the SOC accompanying future 10 CFR 50.55a amendments, which will be published in a FRN.

References to American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI Used in This Report

To aid applicants in the development of their subsequent license renewal applications (SLRAs), the staff has developed a list of aging management programs (AMPs) in the Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report that are based on consistency with the 10-program element criteria defined in Section A.1.2.3 of the Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants (SRP-SLR). Some of the AMPs referenced in the GALL-SLR Report are based entirely or in part on compliance with the requirements of ASME Code Section XI, as endorsed for use through reference in 10 CFR 50.55a. In addition, in some cases, the staff has determined that specific requirements in ASME Code Section XI need to be augmented to ensure adequate aging management consistent with the requirements of the License Renewal Rule. Thus, some of the AMPs in the GALL-SLR Report also provide guidance on augmenting the requirements of ASME Code Section XI. The staff has determined that in most cases the ASME Code Section XI requirements referenced in Table I-1 provide an acceptable basis for managing the effects of aging during the subsequent period of extended operation, except where noted and augmented in the GALL-SLR Report. Therefore, except where noted (see below) and augmented in the GALL-SLR Report, the ASME Code Section XI editions and addenda listed in Table I-1, subject to the modifications and limitations in 10 CFR 50.55a, should be treated as consistent with the GALL-SLR Report, and an applicant need not identify exceptions when using these specific editions and addenda. It should be noted that in some instances, AMPs have been augmented by referencing an edition or addenda beyond that referenced in Table I-1. In these cases, the staff has reviewed the provisions of the later code and has specifically cited the later edition or addenda based on the requirements contained within that version of ASME Code Section XI. In order for an applicant’s program to be consistent with such an AMP, the later edition and addenda should be cited.

An applicant should identify exceptions to the GALL-SLR Report and provide justification when using any ASME Code Section XI edition or addenda not listed in Table I-1 or specifically cited in a GALL-SLR Report AMP. With respect to more recent (beyond those already cited in a GALL-SLR Report AMP) ASME Code Section XI editions and addenda, the NRC will update Table I-1 through either a published revision to the GALL-SLR Report or through the license renewal interim staff guidance process after the staff has evaluated the specific ASME Code Section XI edition or addendum and determined the extent to which it is adequate for license renewal.